• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.21 no.4
• /
• pp.451-463
• /
• 2023

With South Korea increasingly focusing on nuclear energy, the management of spent nuclear fuel has attracted considerable attention in South Korea. This study established a novel procedure for selecting safety-relevant radionuclides for long-term safety assessments of a deep geological repository in South Korea. Statistical evaluations were performed to identify the design basis reference spent nuclear fuels and evaluate the source term for up to one million years. Safety-relevant radionuclides were determined based on the half-life criteria, the projected activities for the design basis reference spent nuclear fuel, and the annual limit of ingestion set by the Nuclear Safety and Security Commission Notification No. 2019-10 without considering their chemical and hydrogeological properties. The proposed process was used to select 56 radionuclides, comprising 27 fission and activation products and 29 actinide nuclides. This study explains first the determination of the design basis reference spent nuclear fuels, followed by a comprehensive discussion on the selection criteria and methodology for safety-relevant radionuclides.

• Nuclear Engineering and Technology
• /
• v.56 no.8
• /
• pp.3188-3198
• /
• 2024

Efficient and secure decommissioning of nuclear facilities demands advanced technologies. In this context, gamma-ray detection and imaging are crucial in identifying radioactive hotspots and monitoring radiation levels. Our study is dedicated to developing a gamma-ray detection system tailored for integration into robotic platforms for nuclear decommissioning, offering a safe and automated solution for this intricate task and ensuring the safety of human operators by mitigating radiation exposure and streamlining hotspot localization. Our approach integrates a Compton camera based 3D reconstruction algorithm with a single Timepix3 detector. This eliminates the need for a second detector and significantly reduces system weight and cost. Additionally, combining a 3D camera with the setup enhances hotspot visualization and interpretation, rendering it an ideal solution for practical nuclear decommissioning applications. In a proof-of-concept measurement utilizing a 137Cs source, our system accurately localized and visualized the source in 3D with an angular error of 1° and estimated the activity with a 3% relative error. This promising result underscores the system's potential for deployment in real-world decommissioning settings. Future endeavors will expand the technology's applications in authentic decommissioning scenarios and optimize its integration with robotic platforms. The outcomes of our study contribute to heightened safety and accuracy for nuclear decommissioning works through the advancement of cost-effective and efficient gamma-ray detection systems.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.13 no.1
• /
• pp.21-34
• /
• 2015

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.

• Economic and Environmental Geology
• /
• v.49 no.4
• /
• pp.301-314
• /
• 2016

Disused Sealed Radioactive Sources (DSRSs) are stored temporally in the centralized storage facility of Korea Radioactive Waste Agency (KORAD) and planned to be disposed in the low- and intermediate-level radioactive waste (LILW) disposal facility in Gyeongju city. In this study, preliminary post-closure safety assessment was performed for DSRSs in order to draw up an optimum disposal plan. Two types of disposal options were considered, i.e. engineered vault type disposal and rock cavern type disposal which were planned to be constructed and operated respectively in LILW disposal facility in Gyeongju city. Assessment end-point was individual effective dose of critical group and calculated by using GoldSim code. In normal scenario, the maximum dose was estimated to be approximately $1{\times}10^{-7}mSv/yr$ for both disposal options. It meant that both options had sufficient safety margin when compared with regulatory limit (0.1 mSv/yr). Otherwise, in well scenario, the maximum dose exceeded regulatory limit of 1 mSv/yr in engineered vault type disposal and the exposure dose was mainly contributed by $^{226}Ra$, $^{210}Pb$ (daughter nuclide of $^{226}Ra$) and $^{237}Np$ (daughter nuclide of $^{241}Am$). For rock cavern type disposal, even though the peak dose satisfied regulatory limit, the exposure doses by $^{14}C$ and $^{237}Np$ were relatively high above 10% of regulatory limit. Therefore, it is necessary to exclude $^{14}C$, $^{226}Ra$ and $^{241}Am$ for two type of disposal options and additional management such as long-term storage and development of disposal container for those radionuclides should be performed before permanent disposal for conservative safety and security.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.3 no.4
• /
• pp.309-317
• /
• 2005

A study on the separation of $^{99}Tc,\;^{94}Nb,\;^{55}Fe,\;^{90}Sr\;and\;^{59/63}Ni$ in various radioactive wastes discharged from nuclear power plants has been performed for a use in their quantification which is indispensible for the evaluation of the radionuclide inventory Ni was recovered along with Ca, Mg, Al, Cr, Ti, Mn, Ce, Na, K, and Cu through the sequential separation procedure of Re(as a surrogate of $^{99}Tc$), Nb, Fe and Sr by anion exchange and Sr-Spec extraction chromatography. In this research, chemical separation of Ni from the co-existing elements was investigated by cation exchange and Ni-Spec extraction chromatography. Precipitation behaviour of Ni and the co-existing elements with dimethylglyoxime(DMG) was investigated in ammonium $citrate/ethanol-H_2O$ and tartaric $acid/acetone-H_2O$ in order to purify separated Ni fractions and to prepare $^{59/63}Ni$ source for the radioactivity measurement using a gas proportional counter. Recovery of Ni separated through ion exchange chromatographic separation procedure was $92.1\%$ with relative standard deviation of $0.9\%$. In addition, recovery of Ni with DMG in the tartaric $acid/acetone-H_2O$ was $85.6\%$ with relative standard deviation of $1.9\%$.

• Journal of Radiation Protection and Research
• /
• v.39 no.4
• /
• pp.159-167
• /
• 2014

In Korea, July 2012, the law as called 'Act on Safety Control of Radioactive Rays Around Living Environment' was implemented to control the consumer product containing Naturally Occurring Radioactive Material (NORM), but, there are no appropriate database and effective dose calculation system. The aim of this study was to develop evaluation technique of the exposure dose with the use of the consumer products containing NORM and to understand the characteristics of the exposed dose according to the radiation type and energy. For the evaluate of exposure dose, the ICRP reference phantom was simulated by the MCNPX code based on Monte Carlo method, and the minimum, medium, maximum energy of alphas, betas, gammas from the representative NORM of Uranium decay series were used as the source term in the simulation. The annual effective doses were calculated by the exposure scenario of the consumer product usage time and position. Short range of the alpha and beta rays are mostly delivered the dose to the skin. On the other hand, the gamma rays mostly delivered the similar dose to all of the organs. The results of the annual effective dose with $1Bq{\cdot}g^{-1}$ radioactive stone-bed and 10% radioactive concentration were employed with the usage time of 7 hours 50 minute per day, the maximum annual effective dose of alphas, betas, gammas were calculated 0.0222, 0.0836, $0.0101mSv{\cdot}y^{-1}$, respectively.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.21 no.1
• /
• pp.70-75
• /
• 2017

Purpose Radioactive iodine thyroid uptake (RAIU) rate is an examination which determines and seeks about general functions of thyroid gland. The size of thyroid gland is normally different between each person, also patients having thyroid diseases have had a variety of size of thyroid gland compared with others. The purpose of this study will investigate about the counting rate which is effected by the geometric factors through the length and volume changes of the source in RAIU rate. Materials and Methods I-131 185 kBq ($5{\mu}Ci$) were placed in a cylindrical phantom of 0.5 cm, 1 cm, 1.5 cm, and 3 cm in diameter, respectively, and saline was added to gradually increase the length by 1 cm in the horizontal and vertical directions to give a change in volume. The source was measured 20 times for 20 seconds from a distance of 25 cm to $364.4keV{\pm}20%$ energy ROI with Captus 3000 thyroid uptake system (Capintec, NJ, USA). Results When the source was located in the transverse direction of the detector, the consequence of one-way ANOVA is that even though the length of source is increased each diameter, there is mostly no significant difference. When the source was located in the longitudinal direction and the counting rate of length 1 cm at all diameter is set to 100%, the average is 92.57% for length 2 cm, 86.1% for 3 cm, 80.69% for 4 cm, 74.82% for 5 cm, and 69.68% at 6 cm. Conclusion According to this study, it is expected that the gap of RAIU rate has been depended on the thickness of thyroid gland as well as the diameter of the beaker. We know that the change of the volume with the increase of the length of the source had less effect on the change of the counting rate. Thus, in order to reduce the error in the measurement of the counting rate with the thyroid uptake rate equipment, an accurate counting rate can be relatively measured if the counting rate which is measured is corrected by thickness or the distance between the thyroid and the thyroid uptake rate equipment is changed.

• Journal of Radiation Protection and Research
• /
• v.22 no.4
• /
• pp.299-307
• /
• 1997

The individual dose equivalent, $H_p$, effective dose, E, and gender specific effective dose, $E^m$ and E$^f$, were evaluated using the male and female phantoms of MIRD type located in the radial gamma radiation field near a point source. The point sources were placed at the distances of 15, 40 and 100 cm in front of the body at different heights. Two radionuclides, $^{137}Cs$ and $^{131}I$, were selected for the illustrative examples. In terms of the gender specific effective doses, $E^f$ is higher than $E^m$ with a few exceptions, e.g. the case where the point source is at the height of reproductive organs, but the differences from the sex- averaged values are not significant enough to justify use of gender specific dose conversion factors for the radial gamma field. The ratios $H_p$/E were in the range of 1 to 3 depending on the source and dosimeter positions when the dosimeter is worn on the front surface of the torso covering from chest to lower abdomen, but varied from 0.34 to 6.5 in extreme cases. When it is assumed that the typical handling procedure of radioactive source material and the typical dosimeter position(on the chest) be respected, the dosimeters calibrated against the broad parallel field appear to provide estimates with acceptable errors for the effective dose of workers exposed to radial broad gamma field around a point source.

• Journal of the Korean Magnetics Society
• /
• v.24 no.6
• /
• pp.186-190
• /
• 2014

To establish the coating conditions for $^{57}Co$, non-radioactive Co ions are dissolved in an acid solution and electroplated on to a copper plate. Then, the thermal diffusion of electroplated Co into a copper matrix was studied to apply a $^{57}Co$ $M{\ddot{o}}ssbauer$ source. Nanocrystalline Co particles were coated on a Cu substrate using DC electro-deposition at a pH of 1.89 to 5 and $20{\sim}30mA/cm^2$. The average grain size was up to 54 nm as the pH increased to 5. The second phase of Co-oxide was formatted as the pH was increased above 4. The diffusion degree was evaluated by mapping using scanning electron microscopy (SEM). The influence of different annealing conditions was investigated. The diffusion depth of Co depends on the annealing temperature and time. The results obtained confirm that the deposited Co diffused almost completely into a copper matrix without substantial loss at an annealing temperature of $900^{\circ}C$ for 2 hours.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.11a
• /
• pp.21-26
• /
• 1996

The identification of radioactive source in a medium with a limited number of external detectors is introduced as an inverse radiation transport problem. This kind of inverse problem is usually ill-posed and severely under-determined, however, its applications are very useful in manu fields including medical diagnosis and nondestructive assay of nuclear materials. Therefore, it is desired to develop efficient and robust solution algorithms. As an approach to solving inverse problems, an artificial neural network is proposed. We develop a modified version of the conventional Hopfield neural network and demonstrate its efficiency and robustness.